مقاله انگلیسی رایگان در مورد انتقال حرارت به فشار آب فوق بحرانی در کانال ها – الزویر 2018

Recent studies on heat transfer to super-critical water (SCW) in tubes, annuli and rod bundles have been reviewed in support of the development of supercritical water-cooled reactors. Experimental investigations are primarily focused on the heat transfer deterioration (HTD) to examine its general behavior, transition boundary and physical mechanisms. Large amount of experimental data were obtained from the experiments supplementing the extensive database previously compiled for fossil fuel-fired power plants. Prediction methods for heat-transfer coefficient were developed from various databases. These methods provide reasonable predictions at normal and enhanced heat-transfer regions, but fail to capture HTD. The upstream effects have not been considered in the prediction methods and may have an impact on local heat transfer, particularly in a channel with a non-uniform axial power profile or with flow/pressure transients. Most numerical studies evaluated the applicability of turbulence models to SCW using the computational fluid dynamics tools. Significant challenges remain in establishing the reliability of the turbulence models and the modeling of buoyancy and turbulent heat flux. Direct numerical simulation and large eddy simulation have been applied in understanding the HTD phenomena. These studies are limited to simple channels over a short axial distance at relatively low Reynolds numbers.

Introduction

Supercritical pressure fossil fuel-fired power plants with water as coolant have been widely adopted to improve the thermal efficiency (currently about 48%) [1, 2]. The use of supercritical pressure water (SCW) in nuclear power plants was explored in the 1960s. Since 2000, there is a renewed interest in developing the Super-Critical Water-cooled Reactor (SCWR) to improve the economic, safety, proliferation resistance and sustainability of the current generation of nuclear systems for commercialized by 2030 [3]. Several conceptual designs have been developed, including the Super-Critical Light-Water Reactor (SCLWR) and the Super-Critical Fast Reactor (SCFR) of Japan [4], High Performance Light-Water Reactor (HPLWR) of Europe [5], Canadian SCWR of Canada [6] and Super-Critical Pressure Vodo-Vodyanoi Energetichesky Reactor (VVER-SKD) of Russia [7]. A collaborative effort has been established for Research and Development (R&D) in support of the development under the Generation-IV International Forum (GIF) [8]. Raising the pressure and temperature above the thermodynamic critical point of water (374 °C, 22.1 MPa) for the SCWRs increases the thermal efficiency from about 33% of the light-water reactors to as high as 45% [9]. This would enhance the fuel utilization and minimize waste stream. Furthermore, phase change of water is not encountered during normal operation at supercritical pressures facilitating the direct transfer of coolant from the reactor outlet to the high-pressure turbine (i.e., direct cycle) eliminating the needs of steam generators (as in pressurized-water reactors) and moisture separators (as in boiling-water reactors). This simplifies considerably the nuclear steam supply system and reduces the reactor plant size and footprint, which leads to significant capital cost saving compared to the current generation of nuclear reactors [3]. Thermal-hydraulics has been identified as one of the critical knowledge areas for the SCWR development [8]. Collaborative R&D are established for heat transfer, hydraulic resistance, stability and critical flow within the GIF-SCWR System. Extensive efforts have been devoted to understanding the heat transfer from channels to SCW, which has a strong impact on the design of core and fuel of the SCWRs.